You are being redirected because this document is part of your ASTM Compass® subscription.
    This document is part of your ASTM Compass® subscription.

    If you are an ASTM Compass Subscriber and this document is part of your subscription, you can access it for free at ASTM Compass

    Simulating the Behavior of Zirconium-Alloy Components in Nuclear Reactors

    Published: 01 January 2002

      Format Pages Price  
    PDF (432K) 17 $25   ADD TO CART
    Complete Source PDF (24M) 882 $435   ADD TO CART

    Cite this document

    X Add email address send
      .RIS For RefWorks, EndNote, ProCite, Reference Manager, Zoteo, and many others.   .DOCX For Microsoft Word


    To prevent failure in nuclear components, one needs to understand the interactions between adjacent materials and the changes in their physical properties during all phases of reactor operation. Three examples from CANDU reactors are described to illustrate the use of simulations that imitate complicated reactor situations. These are: • swelling tests that led to a method for increasing the tolerance of Zircaloy fuel cladding to power ramps • observations of the behavior of leaking cracks in Zr-2.5Nb pressure tubes that provide confidence in the use of leak-before-break as part of the defense against flaw development • contact boiling tests on modifications to the surfaces of Zircaloy calandria tubes that enhance the ability of the heavy water moderator to act as a heat sink after a postulated loss-of-coolant accident


    simulation, zirconium alloys, fuel cladding, pressure tubes, calandria tubes, fuel swelling test, power ramp, leak-before-break, CRACLE, contact conductance, surface modification, heat sink

    Author Information:

    Coleman, CE
    Researcher Emeritus, AECL, Chalk River Laboratories, Chalk River, Ontario

    Committee/Subcommittee: B10.02

    DOI: 10.1520/STP11380S